W3P85-2697, Forwards post-trip Review Requirements,In Response to NRC 851003 SER Requesting Addl Info Re Util 840206 Response to Generic Ltr 83-28,Item 1.1

Forwards post-trip Review Requirements,In Response to NRC 851003 SER Requesting Addl Info Re Util 840206 Response to Generic Ltr 83-28,Item 1.1

ML20133K615
Person / Time
Site:	Waterford
Issue date:	10/15/1985
From:	Cook K LOUISIANA POWER & LIGHT CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
GL-83-28, W3P85-2697, NUDOCS 8510220205
Download: ML20133K615 (14)

Text

o Lo uisiama / ,4e oe-e~oe ex-e , 9 e eexecoe POWER & LIGHT / NEW OALE ANS LOUSANA 701744o0. 9 (5o4) 3e64345 MIDDLE SOUTH uTiuTIES SYSTEM October 15, 1985 s W3P85-2697 A4.05 NQA Director of Nuclear Reactor Regulation Attention: Mr. G.W. Knighton, Chief Licensing Branch No. 3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

Waterford SES Unit 3 Docket No. 50-382 Generic Letter 83-28, Item 1.1 Post-Trip Review

REFERENCES:

(1) W3P84-0297 dated February 6, 1984 (2) NRC letter dated October 3, 1985, Safety Evaluation Report for Waterford 3 Response to Generic Letter 83-28, 1.1 (Post-Trip Review)

Dear Sir:

By Reference (1), LP6L responded to Item 1.1 of Generic Letter 83-28 by supplying a description of the Waterford 3 post-trip review process, including a copy of the post-trip review procedure steps. It is our understanding that your review of the submittal was du e without the benefit of the procedure attachment, resulting in your issuance of additional questions via Reference (2).

To remedy this situation, enclosed please fit.d a copy of the Waterford 3 Post-Trip Review (PTR) requirements. This document, which has been updated since the original Reference (1) submittal, is included as an attachment to procedure OP-10-001 General Plant Operations.

Both the current PTR and the enclosure to Reference (1) address the concerns raised in your Reference (2) letter. Specifically, you note that prior to restart, "an analysis should be conducted to verify that systems which are important to reactor safety have performed as required." The Waterford 3 PTR requires that extensive information as to safety system performance be gathered and assessed prior to restart. See, for example, Parts II-VI of the attached PTR.

The other area of concern in Reference (2) dealt with performance of an independent assessment of the event should review guidelines not be met. The Waterford 3 PTR process requires, in the event that the cause of the reactor '

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Mr. G. W. Knighton W3P85-2697 Page 2 trip is not identified within eight hours, that the STA Coordinator and two members of his staff perform an assessment of the event. The STA Coordinator

. heads a group independent of the Control Room Staff.

We trust that this information is sufficient to resolve your questions on the PTR process for Waterford 3. Please contact Mike Meisner (504-595-2832) should you require additional information.

Yours very truly, 1

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K.W. Cook Nuclear Support & Licensing Manager KWC/MJM/pcl Enclosure cc: B.W. Churchill, W.M. Stevenson, R.D. Martin, J. Wilson, T.A. Flippo, D. Schum a

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POST-TRIP REVIEB PURPOSE The purpose of the Post-Trip Review (PTR) is to identify any malfunction or abnormality and subsequently provide a more detailed investigation and docu-mentation of an unscheduled reactor trip. Also, the PTR provides the necessary information in making the determination that the plant can be restarted safely. A PTR is a prerequisite for a reactor restart following an unsched-uled reactor trip.

REFERENCES NUREG - 1000, Vol. 1 Section 2.2 INPO - OP-211, Post-Trip Reviews .

DEFINITIONS CAUSE - The root initiator of an event (usually an equipment error, proced-ural, or personnel error). When the cause is corrected, the possibility of the event recurring is minimized.

REACTOR TRIP - A manual or automatic insertion of control rods into the-reactor core to interrupt the reactor's ability to sustain a chain reaction.

SEQUENCE OF EVENTS - A hard copy display of the chronological sequence of

! major plant alaras, trips, and actuations, i

i OP-10-001 Revision 4 Attachment 8.19 (1 of 12) 134 e

POST-TRIP REVIEW RESPONSIBILITIES DUTY PLANT MANAGER - The Duty Plant Manager, is responsible for evaluating the recommendations made by the personnel performing the trip investigation and for making the decision for a reactor restart following a Type II Event.

SHIFT SUPERVISOR - The Shift Supervisor is responsible for safety assessment, review and approval of the PTR, and for making the decision for a reactor restart following a Type I Event.

SHIFT TECHNICAL ADVESOR (STA) - The STA is responsible for collecting infor-nation and documenting the information on the PTR. The STA may consult plant personnel for their observation and/or participation in the unscheduled reactor trip event. The STA is also responsible for assisting the shift supervisor in identifying the cause(s) of a reactor trip.

CONTROL ROOM SUPERVISOR - The Control Room Supervisor is responsible for  :

assisting the STA in the reconstruction of the unscheduled reactor trip, if needed. >

STACOORDINATORANDTWOMEMBERSOFHI!iSTAFF-TheSTACOORDINATORandtwo members of his staff are responsible for the review of the PTR if the cause  !

of the reactor trip is not positively identified within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

INSTRUCTIONS - The PTR is a four-step process .The PTR shall not distract  !

the Shift Supervisor, Operating personnel, or STA from their primary responsi-

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bility of maintaining the plant in a safe condition.  !

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I OP-10-001 Revision 4 Attachment 8.19 (2 of 12) f i  !

135 l

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POST-TRIP REVIES i

STEP RESPONSIBILITY

1. Data collection and documentation STA I

2. Trip investigation, event reconstruction SS/CRS/STA and safety assessment i

3. Restart decision Plant Manager or designee (Type II) i Shift Supervisor (Type I)  !

4. Forward the completed PTR to Project Files for record retention as per '

operating instruction 0I-012-000.

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OP-10-001 Revision 4 Attachment 8.19 (3 of 12) 136 r

POST-TRIP REVIEW Date Time Prepared By (STA):

(Name)

NOTE: STA Coordinators to complete Part I through part V. Then submit the package to the Shift Supervisor.

Part I. INITIAL CONDITIONS Reactor Trip:

(Time) (Dote)

Rei<: tor Pave.r  %

1 s

No. cf RCP's (circle)

Turbine Control MAN / AUTO Turbine Bypass Valves MAN / AUTO Which Charging Pump (s) on? A A/B B M W Pump A ON/0FF MW Master Cont A MAN / AUTO MW Speed Cont A MAN / AUTO MW MN Feed Reg. Valve A MAN / AUTO MW Bypass Valve A MAN / AUTO OP-10-001 Revision 4 Attachment 8.19 (4 of 12) 137

POST-TRIP REVIEW (circle)

MW Pump B ON/0FF MW Master Cont B MAN / AUTO MW Speed Cont B MAN / AUTO MW MN Feed Reg. Valve B MAN / AUTO MW Bypass Valve B MAN / AUTO Pressurizer Level Cont MAN / AUTO Pressurizer Press Cont MAN / AUTO

, Pressurizer Spray Cont MAN / AUTO CEA Mode Select Switch Position MAN / AUTO I

0FF-NORMAL STATUS OF ANY SAFETY TRAIN: i RPS SIS CSS CIS EFW MSIS  :

EMERGENCY POWER ANY SAFETY-RELATED TESTING /SURVEILLANCES IN PROGRESS l t

i b

OP-10-001 Revision 4 Attachment 8.19 (5 of 12) 138

, POST-TRIP REVIES

. Part II. PLANT RESPONSE i

Reactor Protection System Type of trip Time Did all CEA's drop? l HP Safety Injection Systes Cause of actuation Time No. of trains  :

LP Safety Injection System Cause of actuation Time '

No. of trains  !

i Containment Spray System ,

cause of actuation Time No. of trains Containment Isolation Systes  ;

Cause of actuation Time  !

i Emergency Feedwater Actuation time .  !

Which pumps started?

Did S/G 1evel' respond?

i Emergen y Power I Actuation time?

Did both diesels start? j Did both diesels load?

Loss of site power? i f

f OP-10-001 Revision 4 Attachment 8.19 (6 of 12) 139

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POST-TRIP REVIEW Safety Injection Tanks actuation time Pressurizer codes time Main Steam codes time Codes resent Yes/No Did Reactor Power Cutback actuate?  ;

What subgroups dropped? '

What event caused' actuation?

Did turbine trip actuate?

Did PZR heater respond normally?

Did PZR level respond normally?

Did PZR spray respond normally?

i Were additional Charging Pumps started?

Time started Time stopped Did pumps start in "AUT0/ MAN"?

Did Steam Generator Level respond normal?

Did Steam Generator Press respond normal?

Was there any AUT0/ MAN station put in " MAN"?  ;

Was there any unplanned radiological release?

OP-10-001 Revision 4 Attachment 8.19 (7 of 12) 140

POST-TRIP REVIE6 Part III. TRANSIENT DATA RCS Pressure max psia sin psia RCS Th Loop 1 max deg. F min deg. F RCS Th Loop 2 max deg. F min deg. F Subcool Margin sax deg F min deg. F SG Pressure .

Loop 1 max psia min psia Loop 2 max psia sin psia SG Level Loop 1 max  % sin  %

Loop 2 max  % sin  % ,

PZR Level max  % sin  % [

Attach parameter recorder plots of:

Reactor Power (Log)

Pressurizer Level 1

Pressurizer Pressure RC Th 1 RC Th 2 i

RC Tc 1 RC Tc 2 M.S. Pressure 1 M.S. Pressure 2 S/G Level 1 S/G Level 2 M.S. Flow 1 M.S. Flow 2  ;

OP-10-001 Revision 4 Attachment 8.19 (8 of 12)

• 141

POST-TRIP REVIES Attachment any E0P attachments performed pursuant to Reactor Trip.

Attach a post-trip primary chemistry analysis.

Attach a copy of the Plant Monitoring Computer (PMC) sequene of events for this trip.

Attach a copy of the PTR program from the PMC.

Attach a copy of pre-trip PR1 chemistry.

Attach a copy of CPC channel trip snapshot (if required as determined by the SS/CRS).

Part IV. PTR SAFETY ASSESSMENT (Circle)

(a) RCS pressure remained above setpoint for Yes / No automatic SI actuation (b) RCS pressure remained below setpoint for PZR Yes / No code safety valve actuation (c) RCS temperature decrease less than 100 deg. F Yes / No per hour (d) Was reactor coolant contained within the Yes / No primary RCS and Quench Tank?

l (e) Indicated PZR level remained on scale Yes / No ;

, (f) Indicated S/G level remained on scale Yes / No l

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OP-10-001 Revision 4 Attachment 8.19 (9 of 12) 142

, POST-TRIP REVIEW ,

. Part V: PLANT PERSONNEL STATEMENTS i

Attach statements from personnel involved with the trip concerning the events that preceded and followed the trip. Each individual should submit a statement concerning the way he renesebers the event.

Example:

Name: Position:

If handwritten statements are prepared, include the plant conditions prior to the trip, your indications that a problem existed, your action as a result of those indications, noted equipment malfunctions or inadequacies, and any identified procedure deficiencies. Also, include any information you consider important to review this unscheduled reactor trip and actions to prevent recurrence.

/

Signature Date / Time Part VI. ANALYSIS AND EVALUATIONS BY SS/CRS/STA 6.1 PROBABLE CAUSE OF TRIP COMMENT OP-10-001 Revision 3 Attachment 8.19 (10.of 12) 143 .

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o POST-TRIP REVIEW 6.2 DESCRIBE ANY UNEXPECTED TRANSIENT BEHAVIOR OR ANY SYSTEMS /OR COMPONENTS INADEQUATE PERFORMANCE.

6.3 IDENTIFY ANY FOLLOW UP ACTION REQUIRED.

6.4 LIST ANY TECHNICAL SPECIFICATIONS THAT WERE EXCEEDED.

Part VII. WAS E CLEAR REGULATORY COMMISSION NOTIFIED IN 1 HOUR?

Part VIII. WERE THE STA COORDINATOR AND AT LEAST 190 MEMBERS OF HIS STAFF NOTIFIED IF THE CAUSE OF THE TRIP.IS NOT IDENTIFIED IN LESS THAN EIGHT HOURS?  !

Part IX. EVENT CONDITION DETERMINATION (BY SS/CRS/STA)

WAS THE CAUSE OF THE REACTOR TRIP, MALFUNCTION TO SAFETY-RELATED AND/OR OTHER IMPORTANT PLANT EQUIPMENT POSITIVELY IDENTIFIED AND CORRECTED AND WERE TECHNICAL SPECIFICATION CONSTRAINTS POSITIVELY IDENTIFIED?

(Circle) Yes No IF ANSWER IS YES, TYPE I EVENT IF ANSWER IS NO, TYPE II EVENT -

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OP-10-001 Revision 4 Attachment 8.19 (11 of 12) 144 l

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Part X. PERMISSION TO START UP t

TYPE I EVENT

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Shift Supervisor's permission to enter mode 2 f i

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Shift Supervisor (Signature) Date Time l f

TYPE II EVENT: '

Duty Plant Manager's permission to enter mode 2 f

Duty Plant Manager (Signature) Date Time i l

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OP-10-001 Revision 4 Attachment 8.19 (12 of 12) i 145 I

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